Fuel-cladding Mechanical Interaction Research Articles

A fine pore-preserved deep neural network for porosity analytics of a high burnup U-10Zr metallic fuel

U-10 wt.% Zr (U-10Zr) metallic fuel is the leading candidate for next-generation sodium-cooled fast reactors. Porosity is one of the most important factors that impacts the performance of U-10Zr metallic fuel. The pores generated by the fission gas accumulation can lead to changes in thermal conductivity, fuel swelling, Fuel-Cladding Chemical Interaction (FCCI) and Fuel-Cladding Mechanical Interaction (FCMI). Therefore, it is crucial to accurately segment and analyze porosity to understand the U-10Zr fuel system to design future fast reactors. To address the above issues, we introduce a workflow to process and analyze multi-source Scanning Electron Microscope (SEM) image data. Moreover, an encoder-decoder-based, deep fully convolutional network is proposed to segment pores accurately by integrating the residual unit and the densely-connected units. Two SEM 250 × field of view image datasets with different formats are utilized to evaluate the new proposed model’s performance. Sufficient comparison results demonstrate that our method quantitatively outperforms two popular deep fully convolutional networks. Furthermore, we conducted experiments on the third SEM 2500 × field of view image dataset, and the transfer learning results show the potential capability to transfer the knowledge from low-magnification images to high-magnification images. Finally, we use a pre-trained network to predict the pores of SEM images in the whole cross-sectional image and obtain quantitative porosity analysis. Our findings will guide the SEM microscopy data collection efficiently, provide a mechanistic understanding of the U-10Zr fuel system and bridge the gap between advanced characterization to fuel system design.

Analysis of the thermal-mechanical behavior of SFR fuel pins during fast unprotected transient overpower accidents using the GERMINAL fuel performance code

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered.Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up.The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

Fuel Performance Analysis of Fast Flux Test Facility MFF-3 and -5 Fuel Pins Using BISON with Post Irradiation Examination Data

Using the BISON fuel-performance code, simulations were conducted of an automated process to read initial and operating conditions from the Pacific Northwest National Laboratory (PNNL) database and reports, which contain metallic-fuel data from the Fast Flux Test Facility (FFTF) MFF Experiments. This work builds on previous modeling efforts involving 1977 EBR-II metallic fuel pins from experiments. Coupling the FFTF PNNL reports to BISON allowed for all 338 pins from MFF-3 and MFF-5 campaigns to be simulated. Each BISON simulation contains unique power and flux histories, axial power and flux profiles, and coolant-channel flow rates. Fission-gas release (FGR), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON and compared to available post-irradiation examination (PIE) data. Cladding profilometry, FGR, and fuel axial swelling simulation results for full-length MFF metallic pins were found to be in agreement with PIE measurements using FFTF physics and models used previously for EBR-II simulations. The main two peaks observed within the cladding profilometry were able to be simulated, with fuel-cladding mechanical interaction (FCMI), fuel-cladding chemical interaction (FCCI), and thermal and irradiation-induced creep being the cause. A U-Pu-Zr hot-pressing model was included in this work to allow pore collapse within the fuel matrix. This allowed better agreement between BISON-simulated cladding profilometry and PIE measurements for the peak caused by FCMI. This work shows that metallic fuel models used to accurately represent fuel performance for smaller EBR-II pins may be used for full-length metallic fuel, such as FFTF MFF assemblies and the Versatile Test Reactor (VTR). As new material models and PIE measurements become available, FFTF MFF assessment cases will be reassessed to further BISON model development.

Analysis of the performance of driver MOX fuel in the MYRRHA reactor under Beam Power Jump transient irradiation conditions

This work focuses on the performance analysis of driver fuel pins under transient irradiation conditions in the MYRRHA reactor, which must be considered besides the normal operation towards the design optimization and licensing of the facility. The considered transient scenario is a Beam Power Jump (BPJ) caused by a trip of the MYRRHA accelerator coupled to the sub-critical reactor core. Based on the outcomes of the analysis of the nominal MYRRHA irradiation - NED 386 (2022) 111581, the impact of this over-power scenario on the pin response is investigated both at the beginning and at the end of irradiation as the most critical moments for the fuel maximum temperature and for the potential fuel-cladding mechanical interaction, respectively. The simulation results are achieved with the TRANSURANUS fuel performance code equipped with advanced models for thermal–mechanical properties of U-Pu mixed-oxide (MOX) fuels, for inert gas behaviour and for the specific mechanical response of DIN 1.4970 cladding. The comparison with design limit criteria adopted highlights the safety of the MYRRHA fuel pins under irradiation, complying with satisfactory margins even considering the occurrence of BPJ transients.

Investigation of swelling behaviors of U-10Zr metallic fuel in the low temperature regime via a cavitational void swelling model

The uranium zirconium (U-Zr) metallic nuclear fuel has long been considered as a promising candidate fuel type for fast breeder reactors. One of the key performance issues of the U-10Zr metallic fuel is its rather strong swelling, which would lead to significant fuel cladding mechanical interaction (FCMI) at relatively high fuel burn-ups. It is therefore imperative to understand the fuel swelling behaviors and the underlying physics mechanisms. Recently published experimental results of U-10Zr fuel swelling behaviors in a low temperature regime (400-600K) showed perceivable swelling of the in-pile irradiated fuel. However, the underlying physics mechanism behind this swelling is not clear. In this work, we performed in-depth analyses on the microstructural features of the irradiated fuel with metallography and conducted a finite-element based simulation on a computational platform, COMSOL, to determine the hydrostatic stress in the irradiated fuel. The spherical objects in the metallography result have been identified as cavities with relatively low fission gas pressure instead of equilibrium gas bubbles. A rate theory code based on the cavitational void swelling model has been developed and the fuel swelling behaviors of U-10Zr fuel in the low temperature regime have been assessed. The simulation results suggest a much lower migration energy for vacancy diffusion in the metallic fuel compared to what is currently used, and a new set of key parameters for the rate theory model were determined.

Preliminary Multi-Physics Performance Analysis and Design Evaluation of UO2 Fuel for LBE-Cooled Subcritical Reactor of China Initiative Accelerator Driven System

The China initiative Accelerator Driven System (CiADS) and the corresponding lead-bismuth eutectic (LBE) cooled subcritical reactor, as the research subject of one of the major national science and technology infrastructure projects, are undertaken by the Institute of Modern Physics-Chinese Academy of Sciences (IMP-CAS). And in the first phase, UO2 fuels will be loaded in the subcritical core to test the coupling technology and achieve a long-term steady operation. A brief description of CiADS subcritical reactor, fuel assembly and fuel element are presented here, and a multi-physics performance analysis and design evaluation of CiADS UO2 fuel are carried out by means of the FUTURE code. FUTURE is a fuel performance analysis code to evaluate the synergy of phenomena occurring in the fuel element and their impact on the fuel design improvement for the liquid metal fast reactor, which was developed jointly by IMP-CAS and Xi’an Jiaotong University (XJTU). In this paper, the FUTURE code was modified and updated focusing on characteristics of CiADS fuels. Relocation and densification models were added. Results of the hottest fuel element, mainly concerning the thermo-mechanical behaviors, are discussed concerning both fuel and cladding performance on the basis of indicative design limits. According to the preliminary design, the CiADS UO2 fuel exhibits good performance, and the main safety parameters are far below the indicative limits. The Fuel Cladding Mechanical Interaction (FCMI) is not very serious, and the permanent cladding strains and Cumulative Damage Fraction (CDF) are small and even negligible thanks to the low level of fuel temperature and corresponding stress. However, some critical issues may still exist, especially on LBE corrosion near the coolant inlet, where protective oxide layers are very thin from BoL to EoL. The modeling is useful for providing feedback to the conceptual design of the CiADS LBE-cooled subcritical reactor and the update of FUTURE code.

Lead-Cooled Fast Reactor Annular UN Fuel Design and Development of Performance Analysis Program

A kind of annular uranium nitride (UN) fuel suitable for lead-cooled fast reactor applications has been designed in this study. The design is directly targeting two main issues of UN fuel: severe swelling and thermal decomposition of UN fuel at high temperatures. A performance analysis program based on FORTRAN programming language has been developed for UN fuel in fast reactors. The program contains heat transfer, fuel stress-strain analysis, cladding stress-strain analysis, fission gas release and fuel-cladding mechanical interaction (FCMI) modules, etc. Extensive code verification has been performed by comparing simulation results obtained with the code and those obtained via the COMSOL Multiphysics platform. Preliminary code validation has been conducted as well by comparing code simulation results with experimental data. The results showed that this program could predict the fuel temperature, stress-strain, and displacement of UN fuel during reactor operation with a reasonable accuracy.

Development of a thermo-mechanical coupling code based on an annular fuel sub-channel code

To simulate the thermal hydraulic performance of the annular fuel more accurately, some mechanical models have been incorporated into the sub-channel code SACAF, including pellet displacement model, gap heat transfer model, fission gas release model and fuel-cladding mechanical interaction model. The results calculated by the developed code are in good agreement with that by FROBA-ANNULAR. The developed code can consider the change of the heat transfer coefficient of the inner and outer gaps, the release of the fission gas release, the inner pressure of the fuel, the stress and strain of the cladding, which cannot be calculated by the conventional sub-channel code. Based on the calculation results, a great change is found in the gap heat transfer coefficient and MDNBR during the operation, which implies that the coupling of the sub-channel code and the mechanical performance models is necessary.

Fuel performance evaluation of annular metallic fuels for an advanced fast reactor concept

The annular metallic fuel design has been considered as a promising solution to relieve the fuel-cladding mechanical interaction (FCMI) caused by significant swelling of metallic fuels at high burnup. In this paper, the fuel performance of annular metallic fuel with HT9 cladding in a novel advanced fast reactor design optimized for high-burnup performance was evaluated. The BISON code was utilized to perform fuel performance simulations using a modified fuel swelling correlation for conservative evaluation. Three representative fuel pins, which feature highest peak burnup (26.8 %FIMA, fission per initial metal atom), highest fuel temperature (645 °C), and highest fast neutron flux (2.68 × 1015 n/cm2·s), were investigated. A series of key fuel performance parameters, such as peak fuel temperature, reduction of fuel central void, and cumulative damage fraction (CDF) of the cladding, were predicted for both upper and lower plenum configurations. It is evident that annular fuel design with a lower smeared density is capable of accommodating the significant swelling of metallic fuel at high burnup. Meanwhile, the adoption of a lower plenum configuration effectively lowers the gas pressure inside the rod and reduces the creep damage accumulation in the HT9 cladding.

Design Study on a Fast Reactor Metal Fuel Attaining Very High Burnup

Innovative fuel design measures to attain a much higher burnup than that obtained using the conventional concept were investigated for a fast reactor (FR) metal fuel. Considering the typical mechanism of metal fuel degradation, three innovative design measures were proposed: (a) a decrease in plenum pressure by adopting the fission gas vent design, (b) prevention of fuel-cladding chemical interaction by lining the cladding inner wall, and (c) mitigation of fuel-cladding mechanical interaction by reducing the fuel smear density. The effects of these design measures on increasing the burnup were analyzed with ALFUS, an irradiation behavior analysis code for FR metal fuels. The ALFUS analysis revealed that a very high burnup of >40 at. % can be attained under the conventional design criteria for securing fuel integrity by applying these innovative measures. Neutronic analysis of a metal fuel core employing these design measures indicated that a high burnup of >40 at. % at the assembly peak can be attained while suppressing the burnup reactivity swing to almost the same level as that of conventional cores with normal burnup through the use of a minor actinide–containing fuel.

Microstructural characterization of high burn-up mixed oxide fast reactor fuel

High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column were observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.

Study on the mechanism of diametral cladding strain and mixed-oxide fuel element breaching in slow-ramp extended overpower transients

Cladding strain caused by fuel/cladding mechanical interaction (FCMI) was evaluated for mixed-oxide fuel elements subjected to 70–90% slow-ramp extended overpower transient tests in the experimental breeder reactor II. Calculated transient-induced cladding strains were correlated with cumulative damage fractions (CDFs) using cladding strength correlations. In a breached high-smeared density solid fuel element with low strength cladding, cladding thermal creep strain was significantly increased to approximately half the transient-induced cladding strain that was considered to be caused by the tertiary creep when the CDF was close to the breach criterion (=1.0), with the remaining strain due to instantaneous plastic deformation. In low-smeared density annular fuel elements, FCMI load was significantly mitigated and resulted in little cladding strain. The CDFs of the annular fuel elements were lower than 0.01 at the end of the overpower transient, indicating a substantial margin to breach. A substantial margin to breach was also maintained in a high-smeared density fuel element with high strength cladding.

Research and Development of Minor Actinide-containing Fuel and Target in a Future Integrated Closed Cycle System

Research and development of minor actinide-containing fuels and targets, i.e., (Pu,Am)O2–MgO, (Pu,Np)O2–MgO, (U,Pu,Np)O2, (U,Pu,Np)N and (Pu,Np,Zr)N, for use in a future integrated closed cycle system that includes fast reactor and accelerator driven sub-critical system is underway. The present statuses of fabrication test and property measurements are given. Design concept of the oxide target is described in detail together with a screening of the support material. A new apparatus for the measurement of mechanical properties at the elevated temperature is installed for use in evaluating the fuel-cladding mechanical interaction. Development histories with future prospects of two types of Np-containing fuels for the fast reactor are mentioned. Preliminary test results for a new nitride target for the accelerator driven sub-critical system are given. Finally, an irradiation test plan in the experimental fast reactor JOYO is briefly described.

Research and Development of Minor Actinide-containing Fuel and Target in a Future Integrated Closed Cycle System

Research and development of minor actinide-containing fuels and targets, i.e., (Pu,Am)O2–MgO, (Pu,Np)O2–MgO, (U,Pu,Np)O2, (U,Pu,Np)N and (Pu,Np,Zr)N, for use in a future integrated closed cycle system that includes fast reactor and accelerator driven sub-critical system is underway. The present statuses of fabrication test and property measurements are given. Design concept of the oxide target is described in detail together with a screening of the support material. A new apparatus for the measurement of mechanical properties at the elevated temperature is installed for use in evaluating the fuel-cladding mechanical interaction. Development histories with future prospects of two types of Np-containing fuels for the fast reactor are mentioned. Preliminary test results for a new nitride target for the accelerator driven sub-critical system are given. Finally, an irradiation test plan in the experimental fast reactor JOYO is briefly described.

Recent improvements in modeling fission gas release and rod deformation on metallic fuel in LMR

In order to analytically investigate irradiation behavior of metallic fast reactor fuels, the authors have developed the ALFUS (ALoyed Fuel Unified Simulator) code. The ALFUS can mechanistically simulate gas release and deformation behavior of the uranium-zirconium alloy fuel. The stress-strain analysis model into which anisotropic strain due to cavitation at the grain and/or phase boundary in the a-uranium phase has been introduced simulates anisotropic deformation of the uranium-zirconium alloy fuel. The models included in the ALFUS are thought reasonable and consistent with knowledge obtained from the irradiation test results. When the fuel slug swells out and comes into contact with the cladding, compressive stress is produced in the slug and decreases volume of the open pore if it has been formed. This is an essential process to keep fuel-cladding mechanical interaction (FCMI) small in case of lower smear density fuel. Analyses with the ALFUS indicate a significant level of FCMI in case of higher than 80% smear density fuel.

Analytical study on deformation and fission gas behavior of metallic fast reactor fuel

In order to analytically investigate irradiation behavior of metallic fast reactor fuels, the authors have developed the ALFUS (ALoyed Fuel Unified Simulator) code. The ALFUS can mechanistically simulate gas release and deformation behavior of the uranium-zirconium alloy fuel. The stress-strain analysis model into which anisotropic strain due to cavitation at the grain and/or phase boundary in the a-uranium phase has been introduced simulates anisotropic deformation of the uranium-zirconium alloy fuel. The models included in the ALFUS are thought reasonable and consistent with knowledge obtained from the irradiation test results. When the fuel slug swells out and comes into contact with the cladding, compressive stress is produced in the slug and decreases volume of the open pore if it has been formed. This is an essential process to keep fuel-cladding mechanical interaction (FCMI) small in case of lower smear density fuel. Analyses with the ALFUS indicate a significant level of FCMI in case of higher than 80% smear density fuel.

高燃焼度（長寿命）高速炉混合酸化物燃料の設計成立性評価

The potential of the Mixed Oxide (MOX) fuel to achieve a high burnup and long life core was studied for four type fuels with the fuel performance code LIFE-IV. Operating conditions, uncertainties and tolerances were varied in calculations and the results were used to establish multi-dimensional response surfaces from the important performance parameters. The Monte Carlo statistical simulation method was applied to determine the effect of statistical tolerance and uncertainty distributions on the fuel performance and to obtain statistical chracteristic of the fuel behavior.(1) From low to medium burnups of 0-100 GWd/t, fission gas swelling dominates Fuel Cladding Mechanical Interaction (FCMI).(2) At high burnup of 150 to 200GWd/t, solid fission product swelling dominates FCMI and fuel performance characteristics are simplified because cladding strain proceeds approximately at a constant rate with burnup.(3) Statistical fuel performance characteristics were evaluated and it was ascertained that MOX has a potential to realize high burnup with 2-σ reliability.(4) The operatioal and fabricational tolerances have little effects on fuel performances.

Fuel transient deformation

Cladding strains resulting from fuel-cladding mechanical interaction in a transient tested fresh fuel pin are assessed against laboratory measurements of high-temperature creep and hot pressing of mixed-oxide fuel pellets. A fuel pin containing nine different fuel sections with varying fuel pellet geometry and density was transiently tested in a MARK IIIA flowing sodium loop under conditions typical of a 1$/s overpower transient. Post-test cladding strain measurements indicated that the largest strains were generated by solid pellets with small gaps while large gap annular pellets generated the smallest strains. High-temperature creep and hot pressing tests on mixed-oxide fuel have been performed at temperatures up to 2600°C. The results indicate that at temperatures above 2300°C, an additional component of creep is operative; while the densification due to hot pressing was considerably less than expected by extrapolating the low-temperature behavior. Both the in- and out-of-reactor data suggest that fuel creep into the center void or hole of a fuel pin is a more effective means of reducing fuel-cladding stress than densification by hot pressing into fuel pellet porosity.

Fuel-Cladding Mechanical Interaction in PCI-Resistant LWR Fuel Designs During Normal Operation and Power Ramping

The fuel cladding mechanical interaction behavior of developmental fuel rods irradiated in the Halden Boiling Water Reactor was evaluated based primarily on rod elongation measurements made during steadystate and power-ramping irradiation. The developmental fuel rod designs were selected based on attributes that were expected to reduce pellet-cladding interaction (PCI) failures during irradiation. Testing results were compared to a nonpressurized reference design with dished-pellet fuel. For the reference rods, there was a relationship between thermal feedback and fuelcladding mechanical interaction during steady-state irradiation. Significant cladding stresses developed in both the axial and hoop directions in the reference rods during power ramping. During power ramping the general cladding stress distribution in fuel rods with annular fuel pellets was primarily axial while cladding stresses in rods with sphere-pac fuel were mostly in the hoop direction. These results are indicative of superior PCI resistance in the annular pellet fuel rod designs when compared to the reference and sphere-pac rods.

Direct observation of fuel-cladding mechanical interaction (FCMI) in mixed-oxide fast reactor fuel pins

Abstract The WSA-1 and WSA-2 fuel pins exhibit experimental evidence of fuel-cladding mechanical interaction (FCMI) as a result of steady-state irradiation. The direct FCMI evidence involves a comparison of local axial and hoop mechanical strain profiles. The determination of the local axial mechanical strain was possible because of the placement of axial hardness marks 12.7 mm apart along a line parallel to the tubing axis spanning the fuel column. The measured cladding local axial and hoop mechanical deformations were the same within experimental error. The experimental results are in contrast to gas pressurized tube data which exhibit no axial mechanical deformation. A substantial amount of indirect evidence further illustrating the influence of FCMI on the cladding mechanical strain profile is also discussed. The conditions leading to steady-state FCMI are: high fuel smear density (i.e. low fuel-cladding gaps and/or high fuel pellet density), thin wall cladding, low cladding swelling and low fission gas pressure.